Neoclassical tearing modes (NTMs) are magnetic islands which increase locally the radial transport and therefore degrade the plasma performance. They are self-sustained by the bootstrap current perturbed by the enhanced radial transport. The confinement degradation is proportional to the island width and to the position of the resonant surface. The q = 2 NTMs are much more detrimental to the confinement than the 3/2 modes due to their larger radii. NTMs are metastable in typical scenarios with β N 1 and in the region where the safety factor is increasing with radius. This is due to the fact that the local perturbed pressure gradient is sufficient to self-sustain an existing magnetic island. The main questions for burning plasmas are whether there is a trigger mechanism which will destabilize NTMs, and what is the best strategy to control/avoid the modes. The latter has to take into account the main aim which is to maximize the Q factor, but also the controllability of the scenario. Standardized and simplified equations are proposed to enable easier prediction of NTM control in burning plasmas from present experimental results. The present expected requirements for NTM control with localized electron cyclotron current drive (ECCD) in ITER are discussed in detail. Other aspects of the above questions are also discussed, in particular the role of partial stabilization of NTMs, the possibility to control NTMs at small size with little ECH power and the differences between controlling NTMs at the resonant surface or controlling the main trigger source, for the standard scenario namely the sawteeth. It is shown that there is no unique best strategy, but several tools are needed to most efficiently reduce the impact of NTMs on burning plasmas.

Abstract. The ITER electron cyclotron (EC) upper port antenna (or launcher) is nearing completion of the detailed design stage and will soon be starting the final build to print design. The main objective of this launcher is to drive current locally to stabilise the NTMs (depositing ECCD inside of the island that forms on either the q=3/2 or 2 rational magnetic flux surfaces) and control the sawtooth instability (deposit ECCD near the q=1 surface). The launcher should be capable of steering the focused beam deposition location to the resonant flux surface over the range in which the q=1, 3/2 and 2 surfaces are expected to be found, for the various plasma equilibria susceptible to the onset of NTMs and sawteeth. The aim of this paper is to provide the design status of the principle components that make up the launcher: port plug, mm-wave system and shield block components. The port plug represents the chamber that provides a rigid support structure that houses the mm-wave and shield blocks. The mm-wave system is comprised of the components used to guide the RF beams through the port plug structure and refocus the beams far into the plasma. The shield block components are used to attenuate the nuclear radiation from the burning plasma, protecting the fragile in-port components and reducing the neutron streaming through the port assembly. The design of these three subsystems is described, in addition, the relevant thermo-mechanical and electro-magnetic analysis are reviewed for the critical design issues.

The ability of the ITER Electron Cyclotron Wave launchers of driving localized current at various plasma locations is analyzed by means of beam tracing codes, looking at extended physics application of ECCD in ITER and at possible synergy between the two launchers. Calculations for an improved design of the Upper launcher, based on four upper ports and front steering mirrors allowing both optimum focusing of the beams and an extended plasma deposition region, show that narrow, high peak current density profiles may be maintained over the radial range 0.4≤ρ p ≤0.9. Calculations for the Equatorial launcher, where the control of the deposition location is achieved by varying the toroidal injection angle β, point out that, due to poor localization and incomplete power absorption at large toroidal angles (β>40 0 ), the power deposition and current drive location by this launcher is limited to ρ≤0.55. Moreover it is shown that performances close to the center can be improved with a poloidal tilt of the low and top front mirrors. The main aim of this study is to provide guidance to the design of both launchers in order to optimize their performances, depending on the physics application.

Abstract.A predictive analysis of the capabilities of the ITER Electron Cyclotron Wave system is presented in this work. Modifications of both the Upper Launcher and the Equatorial Launcher aimed at increasing the potentialities of the system are identified and discussed. A modification of the Upper Launcher has been already incorporated in the updated front steering design called Extended Physics Launcher. By adopting different deposition ranges for the upper and lower steering mirrors, this design offers the flexibility to drive current from the radial region required for the stabilization of neoclassical tearing modes up to that required for sawtooth control, allowing a synergy with the Equatorial Launcher. Here a comparison of the performances of the new design of the Upper Launcher with those of a Dropped Upper Launcher is performed, showing that better performances for both NTM stabilization and sawtooth control may be obtained by lowering the upper port location. An analysis of the EC current driven by the Equatorial Launcher is also presented, showing that adding to the present design the possibility to drive counter-current in addition to the existing co-current capability would significantly increase the flexibility of the system. The behaviour of all launchers at reduced magnetic fields is also discussed.

This paper considers the heating mix of ITER for the two main scenarios. Presently, 73 MW of absorbed power are foreseen in the mix 20/33/20 for ECH, NBI and ICH. Given a sufficient edge stability, Q = 10-the goal of scenario 2-can be reached with 40 MW power irrespective of the heating method but depends sensitively inter alia on the H-mode pedestal temperature, the density profile shape and on the characteristics of impurity transport. ICH preferentially heats the ions and would contribute specifically with Q < 1.5. The success of the Q = 5 steady-state scenario 4 with reduced current requires discharges with improved confinement necessitating weakly or strongly reversed shear, f bs > 0.5, and strong off-axis current drive (CD). The findings presented here are based on revised CD efficiencies γ for ECCD and a detailed benchmark of several CD codes. With ECCD alone, the goals of scenario 4 can hardly be reached. Efficient off-axis CD is only possible with NBI. With beams, inductive discharges with f ni > 0.8 can be maintained for 3000 s. The conclusion of this study is that the present heating mix of ITER is appropriate. It provides the necessary actuators to induce in a flexible way the best possible scenarios. The development risks of NBI at 1 MeV can be reduced by operation at 0.85 MeV.